An approach to combine neutron and ion irradiation data to accelerate material qualification for nuclear reactors
Next-generation nuclear power plants are generally characterized by higher operating
temperatures, increased neutron fluences and energies, and distinct corrosive coolant …
temperatures, increased neutron fluences and energies, and distinct corrosive coolant …
Microstructure response and lead-bismuth eutectic corrosion behavior of 11Cr1Si ferritic/martensitic steel after Au-ion irradiation
Q Chen, F Bai, P Wang, J Yang, C Zhu, W Zhang… - Corrosion …, 2022 - Elsevier
The effect of Au-ion irradiation dose on microstructure evolution and lead-bismuth eutectic
(LBE) corrosion behavior of 11Cr1Si ferritic/martensitic (F/M) steels was investigated. The …
(LBE) corrosion behavior of 11Cr1Si ferritic/martensitic (F/M) steels was investigated. The …
[HTML][HTML] Effect of Au-ions irradiation on microstructure and mechanical properties of FeCrAl coating
W Zhang, J Deng, H Yin, Y Zhao, X Qiu, M Zhou… - Journal of Materials …, 2023 - Elsevier
Six kinds of FeCrAl coatings with different Cr and Al contents were irradiated by 6 MeV Au-
ions with a damage level of 20 dpa. The results showed that the surface roughness and …
ions with a damage level of 20 dpa. The results showed that the surface roughness and …
Analysis of position-dependent cavity parameters in irradiated metals to obtain insight on fundamental defect migration phenomena☆
Motion of point defects is a fundamental process that governs microstructure and properties
of materials. Here, we examine the near-surface and grain boundary cavity swelling depth …
of materials. Here, we examine the near-surface and grain boundary cavity swelling depth …
A quantitative method to determine the region not influenced by injected interstitial and surface effects during void swelling in ion-irradiated metals
We propose and demonstrate a microstructurally-based experimental method to quantitively
determine the depth regions in self-ion-irradiated metals that are affected by the injected …
determine the depth regions in self-ion-irradiated metals that are affected by the injected …
Bridging microscale to macroscale mechanical property measurements of FeCrAl alloys by crystal plasticity modeling
FeCrAl alloys are candidates for accident tolerant fuel cladding of light water reactors. In this
work, a microstructure-and temperature-dependent crystal plasticity model is employed to …
work, a microstructure-and temperature-dependent crystal plasticity model is employed to …
Depth distributions of cavities in advanced ferritic/martensitic and austenitic steels with high helium preimplantation and high damage level
S Jin, H Ma, E Lu, L Zhou, Q Zhang, P Fan, Q Yan… - Materials Today …, 2021 - Elsevier
The cavity depth distributions in advanced ferritic/martensitic and austenitic 316 steels were
characterized after irradiation with Ni ions at 580° C to peaked damage of~ 300 dpa and …
characterized after irradiation with Ni ions at 580° C to peaked damage of~ 300 dpa and …
[HTML][HTML] Homogeneous void nucleation in the presence of supersaturated vacancies and interstitials
L Shao - Journal of Materiomics, 2024 - Elsevier
Homogeneous void nucleation in metals containing arbitrary vacancies and interstitials has
been reexamined, with corrections made to the original work by Katz and Wiedersich. The …
been reexamined, with corrections made to the original work by Katz and Wiedersich. The …
Void swelling of conventional and composition engineered HT9 alloys after high-dose self-ion irradiation
Abstract Ferritic/martensitic (F/M) steels are being considered as potential structural
materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the …
materials for next generation nuclear reactors, and variants of the alloy HT9 are some of the …
Swelling resistance of advanced austenitic alloy A709 and its comparison with 316 stainless steel at high damage levels
Alloy A709 is an austenitic alloy developed for power boiler applications in thermal power
plants and is being considered as a candidate structural material for Generation IV reactors …
plants and is being considered as a candidate structural material for Generation IV reactors …